Refine your search:     
Report No.
 - 
Search Results: Records 1-16 displayed on this page of 16
  • 1

Presentation/Publication Type

Initialising ...

Refine

Journal/Book Title

Initialising ...

Meeting title

Initialising ...

First Author

Initialising ...

Keyword

Initialising ...

Language

Initialising ...

Publication Year

Initialising ...

Held year of conference

Initialising ...

Save select records

Journal Articles

Improvement of plant reliability based on combining of prediction and inspection of crack growth due to intergranular stress corrosion cracking

Uchida, Shunsuke; Chimi, Yasuhiro; Kasahara, Shigeki; Hanawa, Satoshi; Okada, Hidetoshi*; Naito, Masanori*; Kojima, Masayoshi*; Kikura, Hiroshige*; Lister, D. H.*

Nuclear Engineering and Design, 341, p.112 - 123, 2019/01

 Times Cited Count:1 Percentile:58.8(Nuclear Science & Technology)

Improvement of plant reliability based on reliability-centered-maintenance (RCM) is going to be undertaken in NPPs. RCM is supported by risk-based maintenance (RBM). The combination of prediction and inspection is one of the key issues to promote RBM. Early prediction of IGSCC occurrence and its propagation should be confirmed throughout the entire plant systems which should be accomplished by inspections at the target locations followed by timely application of suitable countermeasures. From the inspections, accumulated data will be applied to confirm the accuracy of the code, to tune some uncertainties of the key data for prediction, and then, to increase their accuracy. The synergetic effects of prediction and inspection on application of effective and suitable countermeasures are expected. In the paper, the procedures for the combination of prediction and inspection are introduced.

Journal Articles

An Empirical model for the corrosion of stainless steel in BWR primary coolant

Uchida, Shunsuke*; Hanawa, Satoshi; Naito, Masanori*; Okada, Hidetoshi*; Lister, D. H.*

Corrosion Engineering, Science and Technology, 52(8), p.587 - 595, 2017/10

 Times Cited Count:1 Percentile:87.67(Materials Science, Multidisciplinary)

Based on the relationship among ECP, metal surface conditions, exposure time and other environmental conditions, a model to evaluate the ECP and corrosion rate of steel was developed by coupling a static electrochemical analysis and a dynamic oxide layer growth analysis. Major conclusion obtained on the model are as follows. The effect of H$$_{2}$$O$$_{2}$$ and O$$_{2}$$ concentrations on ECP were successfully explained as the effects of oxide layer growth. Hysteresis of ECP under changes in water chemistry conditions were successfully explained with the model. Decreases in ECP due to neutron exposure were explained well by radiation-induced diffusion in the oxide layers.

Journal Articles

Development of the source term PIRT based on findings during Fukushima Daiichi NPPs accident

Suehiro, Shoichi*; Sugimoto, Jun*; Hidaka, Akihide; Okada, Hidetoshi*; Mizokami, Shinya*; Okamoto, Koji*

Nuclear Engineering and Design, 286, p.163 - 174, 2015/05

 Times Cited Count:12 Percentile:16.19(Nuclear Science & Technology)

The severe accident evaluation committee of AESJ (Atomic Energy Society of Japan) developed the thermal hydraulic PIRT (Phenomena Identification and Ranking Table) and the source term PIRT based on findings during the Fukushima Daiichi NPPs accident. These PIRTs aimed to explore the debris distribution and the current condition in the NPPs with high accuracy and to extract higher priority from the aspect of the sophistication of the analytical technology to predict the severe accident phenomena by the code. The ST PIRT was divided into 3 phases for the time domain and 9 categories for the spatial domain. The 68 phenomena were extracted and the importance from viewpoint of the source term was ranked through brainstorming and discussion. This paper described the developed ST PIRT list and summarized the high ranked phenomena in each phase.

Journal Articles

Detailed analyses of key phenomena in core disruptive accidents of sodium-cooled fast reactors by the COMPASS code

Morita, Koji*; Zhang, S.*; Koshizuka, Seiichi*; Tobita, Yoshiharu; Yamano, Hidemasa; Shirakawa, Noriyuki*; Inoue, Fusao*; Yugo, Hiroaki*; Naito, Masanori*; Okada, Hidetoshi*; et al.

Nuclear Engineering and Design, 241(12), p.4672 - 4681, 2011/12

 Times Cited Count:10 Percentile:31.54(Nuclear Science & Technology)

A five-year research project has been initiated in 2005 to develop a code based on the MPS (Moving Particle Semi-implicit) method for detailed analysis of key phenomena in core disruptive accidents (CDAs) of sodium-cooled fast reactors (SFRs). The code is named COMPASS (Computer Code with Moving Particle Semi-implicit for Reactor Safety Analysis). The key phenomena include (1) fuel pin failure and disruption, (2) molten pool boiling, (3) melt freezing and blockage formation, (4) duct wall failure, (5) low-energy disruptive core motion, (6) debris-bed coolability, (7) metal-fuel pin failure. Validation study of COMPASS is progressing for these key phenomena. In this paper, recent COMPASS results of detailed analyses for the several key phenomena are summarized. The present results demonstrate COMPASS will be useful to understand and clarify the key phenomena of CDAs in SFRs in details.

Journal Articles

COMPASS code development; Validation of multi-physics analysis using particle method for core disruptive accidents in sodium-cooled fast reactors

Koshizuka, Seiichi*; Morita, Koji*; Arima, Tatsumi*; Tobita, Yoshiharu; Yamano, Hidemasa; Ito, Takahiro*; Naito, Masanori*; Shirakawa, Noriyuki*; Okada, Hidetoshi*; Uehara, Yasushi*; et al.

Proceedings of 8th International Topical Meeting on Nuclear Thermal-Hydraulics, Operation and Safety (NUTHOS-8) (CD-ROM), 11 Pages, 2010/10

In this paper, FY2009 results of the COMPASS code development are reported. Validation calculations for melt freezing and blockage formation, eutectic reaction of metal fuel, duct wall failure (thermal-hydraulic analysis), fuel pin failure and disruption and duct wall failure (structural analysis) are shown. Phase diagram calculations, classical and first-principles molecular dynamics were used to investigate physical properties of eutectic reactions: metallic fuel/steel and control rod material/steel. Basic studies for the particle method and SIMMER code calculations supported the COMPASS code development. COMPASS is expected to clarify the basis of experimentally-obtained correlations used in SIMMER. Combination of SIMMER and COMPASS will be useful for safety assessment of CDAs as well as optimization of the core design.

Journal Articles

Detailed analyses of specific phenomena in core disruptive accidents of sodium-cooled fast reactors by the COMPASS code

Morita, Koji*; Zhang, S.*; Arima, Tatsumi*; Koshizuka, Seiichi*; Tobita, Yoshiharu; Yamano, Hidemasa; Ito, Takahiro*; Shirakawa, Noriyuki*; Inoue, Fusao*; Yugo, Hiroaki*; et al.

Proceedings of 18th International Conference on Nuclear Engineering (ICONE-18) (CD-ROM), 9 Pages, 2010/05

A five-year research project has been initiated in 2005 to develop a code based on the MPS (Moving Particle Semi-implicit) method for detailed analysis of specific phenomena in core disruptive accidents (CDAs) of sodium-cooled fast reactors (SFRs). The code is named COMPASS (Computer Code with Moving Particle Semi-implicit for Reactor Safety Analysis). The specific phenomena include (1) fuel pin failure and disruption, (2) molten pool boiling, (3) melt freezing and blockage formation, (4) duct wall failure, (5) low-energy disruptive core motion, (6) debris-bed coolability, and (7) metal-fuel pin failure. Validation study of COMPASS is progressing for these key phenomena. In this paper, recent COMPASS results of detailed analyses for the several specific phenomena are summarized.

Journal Articles

Validation for multi-physics simulation of core disruptive accidents in sodium-cooled fast reactors by COMPASS code

Koshizuka, Seiichi*; Morita, Koji*; Arima, Tatsumi*; Zhang, S.*; Tobita, Yoshiharu; Yamano, Hidemasa; Ito, Takahiro*; Naito, Masanori*; Shirakawa, Noriyuki*; Okada, Hidetoshi*; et al.

Proceedings of 13th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-13) (CD-ROM), 11 Pages, 2009/09

Dispersion and freezing of molten core material was calculated by the COMPASS code to compare with the experimental data of GEYSER. Molten core material flowed up with freezing on the pipe inner surface. As a molten pool behavior, CABRI-TPA2 experiment was analyzed, where a sphere of solid steel was surrounded by solid fuel. Power was injected to cause melting and boiling of the steel sphere. SCARABEE-BE+3 test was analyzed by COMPASS as a validation of failure of duct walls.

Journal Articles

Next generation safety analysis methods for SFRs, 6; SCARABEE BE+3 analysis with SIMMER-III and COMPASS codes featuring duct-wall failure

Uehara, Yasushi*; Shirakawa, Noriyuki*; Naito, Masanori*; Okada, Hidetoshi*; Yamano, Hidemasa; Tobita, Yoshiharu; Yamamoto, Yuichi*; Koshizuka, Seiichi*

Proceedings of 17th International Conference on Nuclear Engineering (ICONE-17) (CD-ROM), 10 Pages, 2009/06

A mesoscopic approach with the COMPASS code is expected to advance the understanding of key phenomena during event progression in core disruptive accidents. In this paper, the overall analysis of SCARABEE-BE+3 test with the SIMMER-III is described as well as the simulation with COMPASS, focusing on the duct wall failure in a small temporal and spatial window cut from the SIMMER-III analysis results.

Journal Articles

Code development for multi-physics and multi-scale analysis of core disruptive accidents in fast reactors using particle methods

Koshizuka, Seiichi*; Morita, Koji*; Arima, Tatsumi*; Zhang, S.*; Tobita, Yoshiharu; Yamano, Hidemasa; Ito, Takahiro*; Shirakawa, Noriyuki*; Naito, Masanori*; Okada, Hidetoshi*; et al.

Proceedings of 16th Pacific Basin Nuclear Conference (PBNC-16) (CD-ROM), 6 Pages, 2008/10

A computer code, named COMPASS, is being developed for various complex phenomena of core disruptive accidents (CDAs) in sodium-cooled fast reactors (SFRs). The COMPASS is designed to analyze multi-physics problems involving thermal hydraulics, structure and phase change, in a unified framework of the MPS (Moving Particle Semi-implicit) method. The project has been carried out by six organizations for five years from FY2005 to FY2009. In this paper, the outcomes of the project in FY2007 are presented. Three validation calculations were completed by following the validation plan: melt freezing and blockage formation, molten pool boiling, and duct wall failure. The COMPASS code development was supported by basic studies of the numerical method, material science for eutectic reaction of the metal fuel, and SIMMER-III analyses.

Journal Articles

Tritium issues toward fusion plants

Konishi, Satoshi; Tobita, Kenji; Nishio, Satoshi; Okada, Hidetoshi*; Kurihara, Ryoichi

Fusion Science and Technology, 41(3), p.817 - 820, 2002/05

no abstracts in English

JAEA Reports

Physics design of advanced steady-state tokamak reactor A-SSTR 2

Nishio, Satoshi; Ushigusa, Kenkichi; Ueda, Shuzo; Polevoi, A.*; Kurita, Genichi; Tobita, Kenji; Kurihara, Ryoichi; Hu, G.; Okada, Hidetoshi*; Murakami, Yoshiki*; et al.

JAERI-Research 2000-029, 105 Pages, 2000/10

JAERI-Research-2000-029.pdf:4.19MB

no abstracts in English

Journal Articles

Control of fusion power in a steady-state tokamak reactor

Kurihara, Ryoichi; Nishio, Satoshi; Ueda, Shuzo; Polevoi, A. R.; Aoki, Isao; Ajima, Toshio*; Okada, Hidetoshi*; Hasegawa, Mitsuru*; Ushigusa, Kenkichi

Journal of Plasma and Fusion Research SERIES, Vol.3, p.553 - 557, 2000/00

no abstracts in English

Oral presentation

Issues associated with source term analysis in accident evaluation and future actions

Maruyama, Yu; Okada, Hidetoshi*; Uchida, Shunsuke*; Hidaka, Akihide

no journal, , 

no abstracts in English

Oral presentation

Necessity of fundamental experiments for evaluating two-phase flow behavior in a RCIC system of the Fukushima Dai-ichi Nuclear Power Plant No.2 unit

Takase, Kazuyuki; Yoshida, Hiroyuki; Okada, Hidetoshi*; Koikari, Soji*; Tsuzuk, Nobuyoshi*

no journal, , 

no abstracts in English

Oral presentation

Risk based strategies for inspection and maintenance of nuclear power plants, 5; An Example for risk based inspection for the events originating IGSCC related causes

Uchida, Shunsuke*; Okada, Hidetoshi*; Naito, Masanori*; Hanawa, Satoshi; Kojima, Masayoshi*; Kikura, Hiroshige*

no journal, , 

no abstracts in English

Oral presentation

Risk based strategies for inspection and maintenance of nuclear power plants, 7; Improvement of plant reliability based on fusion of prediction and inspection on IGSCC

Okada, Hidetoshi*; Uchida, Shunsuke*; Naito, Masanori*; Chimi, Yasuhiro; Hanawa, Satoshi; Kojima, Masayoshi*; Takahashi, Hideharu*; Kikura, Hiroshige*

no journal, , 

no abstracts in English

16 (Records 1-16 displayed on this page)
  • 1